• Corrosion Science and Technology
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• 제18권1호
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• pp.33-38
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• 2019

Alloy 600/182 with excellent mechanical/chemical properties have been utilized for nuclear power plants. Although both alloys are known to have superior corrosion resistance, stress corrosion cracking failure has been an issue in primary water environment of nuclear power plants. Therefore, primary water stress corrosion crack (PWSCC) growth rate tests were conducted to investigate crack growth properties of Alloy 600/182. To investigate PWSCC growth rate, test facilities including water chemistry loop, autoclave, and loading system were constructed. In PWSCC crack growth rate tests, half compact-tension specimens were manufactured. These specimens were then placed inside of the autoclave connected to the loop to provide primary water environment. Tested conditions were set at temperature of $360^{\circ}C$ and pressure of 20MPa. Real time crack growth rates of specimens inside the autoclave were measured by Direct Current potential drop (DCPD) method. To confirm inter-granular (IG) crack as a characteristic of PWSCC, fracture surfaces of tested specimens were observed by SEM. Finally, crack growth rate was derived in a specific stress intensity factor (K) range and similarity with overseas database was identified.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 춘계학술대회
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• pp.274-277
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• 2004

The most important mode of subcritical crack growth is primary water stress corrosion crack, which was the reported mechanism from the root cause analysis of the crack in the bimetallic welds. Stress corrosion crack growth evaluations was carried out for several flaw shapes of both axial and circumferential flaws, using the steady-state stresses including residual stresses. This evaluation considered the possibility of additional flaws in the primary loops of nuclear power plant, even though no such flaws have been identified by Ultrasonic Test. Consequently, Results show that the predicted flaw sizes will determine acceptability for continued service and maintenance.

• Nuclear Engineering and Technology
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• 제51권4호
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• pp.1060-1068
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• 2019

This work aims to establish a model of a primary water stress corrosion crack growth rate of Alloy 690 material for the head penetration nozzles of Korean pressurized water reactors. The test material had an inhomogeneous microstructure with bands of fine-grains and intragranular carbides in the matrix of coarse-grains, which was similar to the archive materials of the head penetration nozzles. The crack growth rate was measured from the strain-hardened materials as a function of the stress intensity factor in simulated primary water at various temperatures and dissolved hydrogen contents. The effects of strain-hardening, temperature, and dissolved hydrogen on the crack growth rate were analyzed independently, and were then introduced as normalizing factors in the crack growth rate model. The crack growth rate model proposed in this work provides a key element of the tools needed to assess the progress of a stress corrosion crack when detected in thick-wall Alloy 690 components in Korean reactors.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 추계학술대회
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• pp.1180-1185
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• 2003

One of the main degradation of steam generator tubes is stress corrosion cracking induced by residual stress. The resulting damages can cause tube bursting or leakage of the primary water which contained radioactivity. Primary water stress corrosion crack occurs at the location of tube/tubesheet hard rolled transition zone. In order to investigate the effect of shot peening on stress corrosion cracking, stress intensity factors are calculated for the crack which is located in the induced residual stress field.

• Nuclear Engineering and Technology
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• 제54권12호
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• pp.4481-4490
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• 2022

In this study, a new rupture disk corrosion test (RDCT) method was developed for real-time detection of stress corrosion cracking (SCC) initiation of Alloy 600 in a primary water environment of pressurized water reactors. In the RDCT method, one side of a disk specimen was exposed to a simulated primary water at high temperature and pressure while the other side was maintained at ambient pressure, inducing a dome-shaped deformation and tensile stress on the specimen. When SCC occurs in the primary water environment, it leads to the specimen rupture or water leakage through the specimen, which can be detected in real-time using a pressure gauge. The tensile stress applied to the disk specimen was calculated using a finite element analysis. The tensile stress was calculated to increase as the specimen thickness decreased. The SCC initiation time of the specimen was evaluated by the RDCT method, from which result it was found that the crack initiation time decreased with the decrease of specimen thickness owing to the increase of applied stress. After the SCC initiation test, many cracks were observed on the specimen surface in an intergranular fracture mode, which is a typical characteristic of SCC in the primary water environment.

• Nuclear Engineering and Technology
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• 제47권7호
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• pp.895-906
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• 2015

Background: Mitigation of primary water stress corrosion cracking (PWSCC) is a significant issue in the nuclear industry. Advanced nickel-based alloys with lower susceptibility have been adopted, although they do not seem to be entirely immune from PWSCC during normal operation. With regard to structural integrity assessments of the relevant components, an accurate evaluation of crack growth rate (CGR) is important. Methods: For the present study, the extended finite element method was adopted from among diverse meshless methods because of its advantages in arbitrary crack analysis. A user-subroutine based on the strain rate damage model was developed and incorporated into the crack growth evaluation. Results: The proposed method was verified by using the well-known Alloy 600 material with a reference CGR curve. The analyzed CGR curve of the alternative Alloy 690 material was then newly estimated by applying the proven method over a practical range of stress intensity factors. Conclusion: Reliable CGR curves were obtained without complex environmental facilities or a high degree of experimental effort. The proposed method may be used to assess the PWSCC resistance of nuclear components subjected to high residual stresses such as those resulting from dissimilar metal welding parts.

• Corrosion Science and Technology
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• 제11권4호
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• pp.141-150
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• 2012

Stress corrosion cracks in Alloy 600 compact tension specimens tested at $325^{\circ}C$ in a simulated primary water environment of pressurized water reactor were analyzed by analytical transmission electron microscopy and secondary ion mass spectroscopy (SIMS). From a fine-probe chemical analysis, oxygen was found on the grain boundary just ahead of the crack tip, and chromium oxides were precipitated on the crack tip and the grain boundary attacked by the oxygen diffusion, leaving a Cr/Fe depletion (or Ni enrichment) zone. The oxide layer inside the crack was revealed to consist of a double (inner and outer) layer. Chromium oxides existed in the inner layer, with NiO and (Ni,Cr) spinels in the outer layer. From the nano-SIMS analysis, oxygen was detected at the locations of intergranular chromium carbides ahead of the crack tip, which means that oxygen diffused into the grain boundary and oxidized the surfaces of the chromium carbides. The intergranular chromium carbide blunted the crack tip, thereby suppressing the crack propagation.

• 대한용접접합학회:학술대회논문집
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• 대한용접접합학회 1998년도 특별강연 및 추계학술발표 개요집
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• pp.288-291
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• 1998

Inconel 600 alloy is used as the material of nuclear steam generator tubing because of its mechanical properties, formability, and corrosion properties. According to reports, the life time of nuclear power plants decreases because of the pitting, intergranular attack, primary water stress corrosion cracking(PWSCC), and intergranular stress corrosion cracking(IGSCC), and denting in the steam generator. The SCC test is very important because of SCC appears in various environment such as solutions, materials, and stress. The C-Rig specimen was made of the steam generator welded sleeve repairing by the pulsed Nd:YAG laser. In the corrosion invironment, corrosion solutions are Primary Water, Caustic, and Sulfate solution and corrosion time is 1624-4877hr. The permitted stress is 30-60ksi.In this C-Ring SCC test is the relationship between corrosion depth, crack and corrosion environment is evaluated. SCC was happens in Sulfate and Corrosion solution but doesn't happen in Primary Water. The corrosion time and stress is very affected by the severely environment of Sulfate or Caustic solution. The microstructure observation indicates that SCC causes interganular failure in the grain boundary of vertical direction.

• 대한기계학회논문집A
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• 제28권6호
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• pp.732-738
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• 2004

One of the main degradation mechanisms in steam generator tubes is stress corrosion cracking induced by residual stress. The resulting damages can cause tube bursting or leakage of the primary water which contains radioactivity. Shot peening technique has been used to prevent stress corrosion crack growth in steam generator tubes. In order to investigate the shot peening effect on stress corrosion cracking stress intensity factors are calculated for the semi-elliptical surface crack which is located in residual stress region. The residual stress distribution in steam generator tubes is obtained from the simple model proposed by Frederick et al.

• 한국압력기기공학회 논문집
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• 제9권1호
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• pp.56-61
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• 2013

This paper presents the results of integrity assessment for the cracks happened in reactor vessel upper head penetration nozzles. The crack morphology for a boat sample from crack area was analyzed through microscope. The stress condition including weld residual stress around crack was analyzed using finite element analysis. From the results of crack morphology and stress condition, the crack was concluded as primary water stress corrosion cracking. The integrity of the cracked nozzle was assessed by the methodology provided in ASME Section XI. According to the assessment results, the remaining life of the cracked nozzle was 1.43 yrs. and the plant decided to repair it.